A Monte Carlo code for the BM 704 computer was written to study the transport of neutrons in a uniform heterogeneous lattice of cylindrical fuel assemblies. Models of the geometric and physical processes are used to obtain the neutron age and migration area; the flux as a function of position, energy, and direction; and absorption data from which thermal utilization and the multiplication constant k may be calculated. (auth)

>The excitation by nuclear collisions of the acoustic modes of a metal hydride crystal was investigated, using a model of the crystal based on experiments on ZrH/sub 2/, but slightly more general. It is found that these modes contribute little tc neutron moderation in ZrH/sub 2/. In the course of the discussion, a generalized form of the Wilkins equation, which determines the spectrum of neutrons thermalizing in a heavy moderator, is developed, applicable when the scattering cross section varies with energy. (auth)

>The proposed flowsheets for reprocessing of nonproduction fuels include centrifugal separation of particulate matter from various dissolver effluent solutions. The settling characteristics of process solids were determined in water and in cold process solutions. Uranium dioxide particles will be recovered from Zirflex and Sulfex cladding waste solutions, and core-dissolver solutions will be centrifuged for removal of ZrO/sub 2/, metallic slimes, siliceous matter, and uranium-bearing solids. (W.L.H.)

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A discussion of properties of operator solutions, of relations between operator solutions, and of the class of all operator solutions is given. Solutions of inhomogeneous boundary-value problems by Green operators are discussed. The connections between operator solutions of scalar and vector problems are studied. (C.J.G.)

Work in the division is summarized in the areas of theoretical studies, model magnet studies, ring magnet vacuum chamber, vacuum pumping system, ring magnet power supply, radio-frequency system, injection system, theoretical studies on radial motion through the linac, outgassing, and ferrite bonding. (For preceding period see ANL-5956.) (W.D.M.)

In view of the more-or-less eminent conversion of at least some of the older Hanford reactors to minimum process tubes it has been requested the feasibility and economics of increasing the process channel size by overboring be studied. This report is concerned with the feasibility from an operational physics standpoint of either raising reactor power levels with present aluminum process tubes and redesigned fuel elements raising reactor power levels with zirconium replacement tubes of current outside diameter, or raising power levels with either zirconium or aluminum tubes of 200 mils greater outside diameter. This report is of a survey nature only, and the data contained herein should be considered in that light.

A 1.6% enriched U rod was irradiated in 105 DR to 450 MWD/T, using a NaK-filled Al capsule. The capsule was discharged in November 1958 and examination begun. Results showed that the NaK in the capsule was trapped in the expansion chambers, leaving part of the fuel rod uncovered, causing gross heating on one side and failure of the rod, preventing the desired data from being obtained.

The operational hazards associated with the use of an isotope-fueled auxiliary power unit for a satellite mission are described. The effects of missile about on the generator are discussed. The generator design is described, and the properties of the various fuel forms are investigated. The characteristics of the fuel capsules and the provisions for biological shielding are also described. Integration of the generator into a typical missile system is discussed. Hazards and procedures of transporting and handling the fuel cores from fabrication to launchlng are considered. Aborted missions are defined, and the forces acting on the generator during abort are described. (W.D.M.)

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A series of hydrodynamic tests was conducted on a 1: 12 scale mcdel of the Yankee reactor vessel and core with four, three, and two loop operation. There was no depression of flow in the center of the core due tc the control rod shrouds or shroud support plate. With four loops operating, flow distribution tests at the core inlet indicated that for all operating conditions the highest velocities occurred in the center region of the core and were a maximum of 18% higher than the average velocities. The lowest velocities were in the region near the core baffle and were a maximum of 18% lower than the average velocities. The effective flow starvation at the points of high flux were less than the 7% design value. Mixing studies indicated that no inlet position of the core received more than 55% of its total now from one loop. The flow distribution did not change appreciably for four, three, or two loop operation. The measured pressure drop across the vessel correlated within 6% of the predicted value. (auth)

A comparison of the creep properties of l5% cold-worked Zircaloy-2 and of annealed Zircaloy-2 is being made. In studies to develop a fuel element leak detector which removes fission products from reactor-coolant streams, experimental work included the deternination of the exchangeabiltty of Br/sup 82/ and AgBr. determination of gross fission-product retention by AgBr columns, and studies to determine possible methods of reducing gross fissionproduet contaniination of AgBr columns. A thermalneutron-flux monitoring system is being developed for the Hanfordreaciors in the development of corrosionresistant welding alloys for use with Hastelloy F, twelve experimental Ni-base alloys have been prepared. Aluminum-35 wt.% U alloys containing small additions of Sn or Zr are being evaluated on the basis of casting and fabricating characteristics mechanical pioperties, and corrosion resistance in 200 ction prod- C water. Data are reported on the effect of fast neutron reactions on the activation analysis of cement ant) cement raw materials. The investigation of radiation- induced free radicals and grafting of polyethacrylates was continued The studies concerned with stabilizing UO/sub 2/ by additions of Ja/sub 2/O/sub 3/ or Y/sub 2/ O/sub 3/ plus CaO wer e continued An investigation is being conducted to determine the effects of high pressure and high …

lattice parameters and space-dependent neutron spectra. The method follows neutron histories through many generations, the total number depending on the statistical uncertainty desired. The major differences in the proposed Monte Carlo scheme and previous Monte Carlo calculations are enumerated. The use of neutron weights is introduced into the routine for the calculation of the fast fission factor. The calculation of the resonance escape probability employs the use of doubling surfaces in the moderator. In computing the thermal utilization, the thermal neutron spectrum is not assumed to be in the usual Wigner-Wilkins form but is calculated for every region. The effects of spectral hardening and epithermal fission are included. The neutron flux is calculated according to the definition that the neutron flux is proportional to the sum of the path lenthhs per cubic centimeter. The proposed method is applicable to computing thermal neutron spectra in both homogeneous and heterogeneous lattices. (C.J.G.)

Surveys of radioactivity in the Clinch and Tennessee rivers during 1954 through 1958 are summarized. It is concluded that no immediate hazard exists due to the reconcentration of radioactive materials in downstream bottom sediments, However, if the amount of radioactivity in the bottom sediment continues to increase for the next few years, it will be necessary to re-evaluate our present waste disposal policy in order to further restrict the release of ralioactive wastes to the Clinch River. The most probable effect of the radioactive sediment on industry would be an increased background counting rate if sand from the river bottom were used in making concrete for the construction of counting rooms of instrument laboratories. The problem ofthe radioactivity in solution in the river water would have to be considered before using the downstream water as process water in the manufacture of film emulsions or other photographic materials, (auth)

A Bausch and Lomb metallograph that had been adapted so that it could be operated remotely was modified further to facilitate its operation. The instrument has performed satisfactorily in the examination of highly radioactive materials behind the heavy shielding of a high level cell. (auth)

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A study was conducted to compare the estimated dissolution rate of Zircaloy-2 clad annular fuel elements on the fully expesed metal with that in the annuli. Results indicate that decladding of these elements should proceed uniformly on all surfaces. Heat balance data from the study are also reported. (J.R.D.)

Numerous types of high temperature ceramic fuel elements for the Pebble Bed Reactor are being evaluated. Specimens are 1 1/2 in. diameter uranium graphite spheres with external coatings such as silicon carbide or pyrolytically deposited high density graphite and fuel particle coatings such as alumina. Low fission product leakage rates at high temperatures have been observed for some of these coatings. High-level irradiation has given no visible evidence of radiation damage to either the silicon carbide coating or the coating-graphite bond. (auth)

This document details the activities of the Fuels Preparation Department during the month of October 1959. (FI)

A quarter-scale flow model, using air as a working fluid, was used to obtain design data for the PEGCPR program. A design for the core-support cylinder, to provide optimum mixing and core-flow distribution, was developed, following which experimentul studies of core-flow distribution, mixing of flow from the two inlets, flow patterns in plenum spaces, flow patterns in the thermal- shield-coolant passage, and pressure drops throughout the model were carried out. Results obtained in the model were then converted to values applicable to helium flow in the prototype. (auth)

During the course of power run 14 on the Sodium Reactor Experiment (SRE) at low power, the temperature difference among various fuel channels was found to be undesirably high Normal operating practices did not succeed in reducing this temperature difference to acceptable values and on July 26, 1959, the run was terminated. A series of fuel element inspections was begun to ascertain the cause of these circumstances, and several fuel elements were discovered to have suffered substantial damage. On July 29, 1959, an Ad Hoc Committee was appointed by Atomics Intennational to assist in the analysis of the existing situation in the reactor and the determination of its origin. During the three-month period since the termination of power run 14, there has been a very active program of investigation. The data accumulated duning the operation of the SRE have been re- examined and evaluated. Metallurgical examination was made of a few samples of the fuel and other components of the reactor where possible. Some chemical analysis was made of the coolant and its contaminants. Radiochemical analyses have been made of the coolant and gaseous activity. Reactivity effects were investigated. Scme experimental programs were initiated to examine mechanisms of damage and …

The mean linear and cubical coefficients of thermal expansion of eight commercial samples of graphite were determined for temperature intervals between 80 and 2000 deg F. The linear thermal expansion was measured with an automatic recording dilatometer using a rod-shaped specimen 2 in. long and 1/4 in. across. The specimen was heated in an atmosphere of helium. The results were in good agreement with those of Currie, Hamister, and MacPherson. The mean linear coefficient was found to increase with temperature. For the samples studied, the mean linear coefficients from 80 to 2000 deg F were 1.50 to 2.34 x 10/sup -6// deg F parallel and 2.26 to 3.45 x 10/sup -6// deg F perpendicular to the grain and were found to vary linearly with the electrical resistivity measured at 32 deg F. (auth)